The following pertains to the nuclear reactor arts, nuclear power arts, nuclear reactor safety arts, and related arts.
Existing nuclear power plants are typically light water thermal nuclear reactors of the boiling water reactor (BWR) or pressurized water reactor (PWR) designs. In such a reactor, a nuclear reactor core comprising fissile material (typically a uranium compound such as UO2 enriched in fissile 235U) is disposed in coolant (purified water) contained at an operational pressure and temperature in a reactor pressure vessel. A nuclear chain reaction involving fission of the fissile 235U generates heat in the nuclear reactor core which is transferred to the coolant. In a BWR design, the heat directly converts coolant to steam, and steam separator/dryer hardware contained in the reactor pressure vessel generates steam that is output via large-diameter piping to a turbine to generate electricity (in a nuclear power plant setting; more generally the output steam is used to perform other useful work). The condensed coolant from the turbine is fed back into the BWR pressure vessel via additional large-diameter piping. In a PWR design, the primary coolant remains in a liquid state (e.g. subcooled) and is piped via large-diameter piping to an external steam generator where heat from the (primary) reactor coolant converts (separate secondary) coolant to steam that in turn drives the turbine. The condensed coolant from the steam generator is fed back into the PWR pressure vessel via additional large-diameter piping.
Safe operation of such reactors includes providing protection against radiological release to the environment. To this end, it is known to surround the nuclear reactor with a radiological containment structure typically constructed of steel and/or steel-reinforced concrete, and to implement safety systems, with redundancy, to remediate events in which reactor operation moves outside of a design envelope. One class of events is a loss of coolant accident (LOCA), in which reactor coolant escapes from a reactor pressure vessel break or, more commonly, from a break in a large-diameter pipe that connects with the reactor pressure vessel at a vessel penetration. A LOCA break which occurs between the vessel penetration and a closest pipe valve is particularly problematic, since reactor coolant loss from such a break continues even after the pipe valve is closed.
A known solution is to provide an integral isolation valve (IIV) at the vessel penetration. An IIV comprises a valve built into a flange that connects with the pressure vessel. Since the IIV is integrated directly into the vessel penetration, closing the IIV ensures stoppage of reactor coolant loss at the LOCA break.
A disadvantage of using IIV's to protect against LOCA events is that pneumatic, hydraulic, or electric control lines are needed to operate the IIV's, and these control lines are run up to the reactor pressure vessel so that they are exposed to heat and radiation flux generated by the operating nuclear reactor. It has been contemplated to employ wireless valve control, but this introduces its own set of problems. The wireless receiver must be built into the IIV and hence is exposed to high temperature and radiation fluxes, and the intangible nature of the wireless communication can make it difficult to detect problems in the valve control system.
An additional disadvantage of using IIV's to protect against LOCA events is that the valve actuator control can be complex, entailing detection of a LOCA condition based on reactor pressure, coolant level, or the like, and operating the IIV's in accordance with the detected reactor condition. The operation is not straightforward, because the response may require keeping some IIV's open and other IIV's closed.